IAEA-TECDOC-1343
Spent fuel performance assessment and research
Final report of a Co-ordinated Research Project on Spent Fuel Performance Assessment and Research (SPAR) 1997–2001
March2003
The originating Section of this publication in the IAEA was:
Nuclear Fuel CycleandMaterials Section International Atomic Energy Agency
Wagramer Strasse 5 P.O. Box 100 A-1400 Vienna, Austria
SPENT FUEL PERFORMANCE ASSESSMENT AND RESEARCH IAEA, VIENNA, 2003
IAEA-TECDOC-1343 ISBN92–0–102703–6
ISSN 1011–4289
© IAEA, 2003
Printed by the IAEA in Austria March2003
FOREWORD
As of the beginning of 2002, more than 150 000 Mg U spent fuel is currently being stored in various storage facilities. Most of this fuel is under water, but dry storage is becoming a widely used technology with more than 9000 Mg U currently stored in various countries. In the future spent fuel storage capacity needs will continue to increase and some of these fuels will have to be stored for 50 years or more, before their reprocessing or final disposal takes place. As a result, interim storage for long periods of time will continue to be a key technology for all Member States.
A number of countries are planning or have already initiated research programmes on spent fuel storage performance, or are developing a high confidence database which evaluates the storage of spent fuel for extremely long periods of time, to determine their consequence on disposal.
Spent fuel storage technology (particularly dry storage) is undergoing a rapid evolution, new fuel and material design changes are coming on stream and target burnups are steadily increasing. These improvements will ask for new studies and potential adaptations of the storage technologies currently used.
As a consequence, the IAEA initiated in 1997 a Co-ordinated Research Project on Spent Fuel Performance Assessment and Research (SPAR) to collect and exchange spent fuel storage experience of the participating countries, to build a comprehensive international database supporting the licensing of present and future technologies; to carry out research work which will evaluate and justify the storage of spent fuel for extremely long periods of time (more than 50 years); and to assist in defining how the requirements for spent fuel storage and for the whole back end of the fuel cycle are connected.
The present publication is based on results obtained in the participating countries. The draft manuscript was prepared and discussed during the last research co-ordination meeting, held in Cordoba, Spain, 1–5 October 2001, and finalized at a consultants meeting in February 2002.
The report provides an overview of technical issues related to spent fuel wet and dry storage and summarizes the objectives and major findings of research, carried out within the framework of the CRP.
The IAEA gratefully acknowledges the contribution of the CRP participants and the consultants who participated in the drafting and review of the report. The IAEA staff member responsible for this publication was H.P. Dyck of the Division of Nuclear Fuel Cycle and Waste Technology.
EDITORIAL NOTE
The use of particular designations of countries or territories does not imply any judgement by the publisher, the IAEA, as to the legal status of such countries or territories, of their authorities and institutions or of the delimitation of their boundaries.
The mention of names of specific companies or products (whether or not indicated as registered) does not imply any intention to infringe proprietary rights, nor should it be construed as an endorsement or recommendation on the part of the IAEA.
CONTENTS
1. INTRODUCTION AND OBJECTIVES... 1
2. HISTORY OF THE BEFAST AND SPAR CO-ORDINATED RESEARCH PROJECTS ... 2
3. WET STORAGE ... 4
3.1. Spent fuel storage experience... 4
3.1.1. General performance... 4
3.1.2. Wet storage experience in the participating countries ... 5
3.2. Pool water chemistry... 9
3.2.1. Water quality control ... 14
3.2.2. Sources of pool water activity... 15
3.3. Inactive content ... 16
3.3.1. Sources of pool water particulates ... 17
3.3.2. Water treatment... 17
3.3.3. Water temperature... 18
3.4. Pool components and materials... 18
3.4.1. Pool lining... 18
3.4.2. Storage racks... 19
4. DRY STORAGE ... 19
4.1. Introduction ... 19
4.2. Dry storage status and experience... 21
4.3. Research activities related to dry storage... 26
5. SOURCE TERMS ... 30
5.1. Subcriticality ... 31
5.2. Thermal calculations ... 32
5.3. Radiation sources ... 32
5.4. Aspects associated with high burnup fuel ... 33
5.5. Aspects associated with burnup credit ... 35
6. FUEL INTEGRITY ... 38
6.1. Definition of fuel integrity ... 38
6.1.1. Loss of fuel cladding integrity ... 41
6.1.2. Loss of fuel assembly lifting features ... 41
6.1.3. Geometrical rearrangement of a fuel assembly ... 41
6.2. Determination of spent fuel integrity ... 41
6.2.1. At reactor detection systems ... 41
6.2.2. Away from reactor detection systems... 44
6.3. Criteria... 45
6.3.1. Utilities... 45
6.3.2. Transportation ... 45
6.3.3. AFR facilities... 45
7. FUEL ASSEMBLY DEGRADATION MECHANISMS IN DRY AND WET STORAGE ... 45
7.1. Degradation mechanisms of fuel assembly components... 45
7.1.1. LWR... 45
7.1.2. AGR ... 46
7.1.3. Magnox ... 47
7.1.4. CANDU ... 47
7.1.5. RMBK andWWER-440/1000... 49
7.2. Degradation mechanisms affecting the fuel cladding ... 49
7.2.1. LWR... 49
7.2.2. Wet storage ... 50
7.2.3. Dry storage... 51
7.2.4. AGR ... 57
7.2.5. Magnox ... 57
7.2.6. CANDU fuel degradation under dry storage conditions... 58
7.2.7. RBMK and WWER-440/1000... 62
7.3. Transition modes between different storage conditions... 62
7.3.1. Transition from wet to dry storage... 62
7.3.2. Dry-to-wet transition... 63
8. BEHAVIOUR OF STORAGE FACILITY COMPONENTS... 63
8.1. Behaviour of metallic components in wet and dry storage ... 63
8.1.1. Stainless steel ... 64
8.1.2. Aluminium alloys... 65
8.1.3. Carbon steels... 66
8.1.4. Cast iron... 66
8.2. Behaviour of reinforced concrete... 66
8.2.1. General properties... 66
8.2.2. Concrete used for wet storage... 68
8.2.3. Concrete used for dry storage... 69
9. ISSUES RELATED TO LONG-TERM STORAGE ... 74
9.1. Extended storage... 74
9.1.1. Extended storage trends ... 74
9.1.2. Review of extended storage projects in the various countries... 75
9.1.3. Data collection and documentation... 78
9.2. Handling of spent fuel after long-term storage ... 79
9.3. Encapsulation/packaging of spent fuel for final disposal... 79
9.3.1. Facilities for spent fuel final packaging... 79
9.3.2. Studies on techniques to support the final disposal of spent fuel ... 86
9.4. Potential loss of inert conditions in dry storage of spent fuel ... 89
9.4.1. Potential temperatures for loss of inert conditions for various fuel types and storage facilities ... 89
9.4.2. Air ingress ... 89
9.4.3. Water ingress... 91
10. MONITORING TECHNOLOGIES AND TECHNIQUES ... 91
10.1. Techniques employed... 91
10.1.1. Visual inspection (including closed circuit TV)... 92
10.1.2. Liquid sampling... 92
10.1.3. Fission gas sampling... 92
10.1.4. Pressure monitoring... 92
10.1.5. Gas sampling ... 94
10.1.6. Radiation detection instrumentation... 94
10.1.7. Temperature measurement ... 94
10.1.8. In-Pool fuel integrity inspection systems ... 94
10.1.9. Burnup monitors... 94
10.1.10. Ultrasonic test ... 94
10.1.11. Corrosion sensors ... 94
10.2. Examples of monitoring... 95
10.2.1. Korean CANDU silo monitoring program ... 95
10.2.2. Ontario Power Generation Used Fuel Dry Storage (UFDS) Monitoring Program ... 95
10.2.3. Inspection of leaking fuel in Sweden ... 96
10.2.4. Sipping of transport flasks, Sweden ... 96
10.3. Regulatory requirements and future demands... 96
ANNEX: COUNTRY REPORTS ON RESEARCH PROJECTS WITHIN THE SPAR CRP... 97
ABBREVIATIONS... 115
REFERENCES... 117
CONTRIBUTORS TO DRAFTING AND REVIEW ... 123
1. INTRODUCTION AND OBJECTIVES
Several options for the ultimate management of spent fuel discharged from nuclear power plants are being implemented or are under consideration. At the present time, the two leading options are:
Direct disposal of the spent fuel in a geological repository (once-through cycle), and Reprocessing of the spent fuel, recycling of the reprocessed plutonium and uranium, and
final disposal of the wastes from the reprocessing operations.
However, delays in opening geological repositories in most countries or in implementing reprocessing in some countries mean that increased spent fuel storage capacity in combination with longer storage durations will be needed over the foreseeable future. Worldwide, the spent fuel production rate, now at about 10 500 t HM, is expected to increase to ~11 500 t HM by 2010. Because less than a third of the fuel inventory is reprocessed, ~8000 t HM/year on average will need to be placed in interim storage facilities.
Presently, spent fuel is mostly stored wet in at-reactor (AR) or at away-from-reactor (AFR) facilities. Because many AR pools are approaching their full capacity, even after extensive re- racking, storage of spent fuel under dry and inert atmosphere is being used increasingly.
The situation is further complicated by today’s reliance on higher enrichment, higher burnup fuels as well as on mixed-oxide fuel, to generate electricity at a competitive cost. Given the much higher decay heat levels from these fuels, wet storage will remain the preferred approach for interim storage during the first decade after discharge. After sufficient decay (“cooling”), and especially for long storage durations (up to 300 years under some scenarios), dry storage under inert conditions or in air become the preferred alternative, given the passive nature of dry storage systems.
This publication is the last report on the IAEA Spent Fuel Performance Assessment and Research (SPAR) co-ordinated research project (CRP) — a continuation of the previous BEFAST-I to -III. Eleven participants from Canada, France, Germany, Hungary, Japan, the Republic of Korea, the Russian Federation, Spain, the United Kingdom, and the United States of America participated in the programme. In addition, a participant from Sweden contributed to the programme as well.
The overall objectives of this CRP were:
1. To carry out research work to evaluate the technical basis for storing spent fuel for extended periods of time, i.e. more than 50 years.
2. To evaluate and exchange data on spent fuel storage research and experience among the participating countries to build a comprehensive, international database supporting the licensing of present and future technologies.
3. To assist in defining how the requirements for spent fuel storage and the whole fuel cycle back-end are connected.
4. To exchange operating experience in spent fuel storage.
The specific objective of the CRP was to identify materials issues in long-term storage facilities.
2. HISTORY OF THE BEFAST AND SPAR CO-ORDINATED RESEARCH PROJECTS
Extended spent fuel storage is, and will remain, an important activity for all countries with nuclear power programmes because fuel after its discharge from the reactor is required to be stored before reprocessing or final disposal. The storage period is highly dependent upon the individual national strategies to close the nuclear fuel cycle, but generic questions related to spent fuel storage are common to all nuclear programmes.
The first phase of the BEFAST project (1981–1986) involved 12 organizations from 11 countries: Austria, Canada, the former Czechoslovakia, Finland, the former Federal Republic of Germany (FRG), the former German Democratic Republic (GDR), Hungary, Japan, Sweden, the USA and the former USSR. A subsequent project, BEFAST-II, Behaviour of Spent Fuel and Storage Facility Components during Long-term Storage, implemented during the years 1986–1991, involved organizations from 12 countries: Argentina, Canada, FRG, Finland, GDR, Hungary, Italy, the Republic of Korea, Japan, the UK, the USA and the USSR.
BEFAST-III, implemented during the years 1991–1996 involved 15 organizations from 12 countries: Canada, Finland, France, Germany, Hungary, Japan, the Republic of Korea, the Russian Federation, Slovakia, Spain, the UK and the USA. There was also an observer from Sweden.
During the three co-ordinated research projects (CRPs) the participating countries contributed their R&D results on fundamental questions of spent fuel storage. The reports of the CRPs have been published as IAEA-TECDOCs [1–3].
Towards the end of the BEFAST-III project, it became apparent that the R&D component of the project was decreasing steadily; more emphasis was being placed on the operation and implementation of storage technology. The storage technology (particularly dry storage) was undergoing a rapid evolution: new fuel and material design changes were coming on stream and target burnup were steadily increasing. With the increased burnup came higher fission gas and fission product inventories, increased sheath (cladding) strains and increased cladding hydriding and oxidation. Because of all the new parameters that have surfaced during the course of BEFAST-III, a new project SPAR was proposed to address the effects of these new parameters on long-term storage and to determine their consequences on disposal.
The first phase of the IAEA Spent Fuel Performance Assessment and Research (SPAR) CRP (1997–2002) involved 11 organisations from 10 countries: Canada (2), France, Germany, Hungary, Japan, Republic of Korea, Russian Federation, Spain, Sweden, UK, and USA.
Sweden participated in the project as an observer.
Three Research Co-ordination Meetings were held during the course of the SPAR CRP: the first in April 1998 in Washington, D.C.; the second in May 2000 in Ahaus, Germany and the third in October 2001 in Cordoba, Spain.
Major topics for both wet and dry storage during all four CRPs are summarised in Table I. It records detailed objectives and the shift in emphasis during the various phases of the Programmes between 1981 and 2002.
Table I. Research Subjects in the CRPs
Long-term Behaviour B-I B-II B-III SPAR Surveillance B-I B-II B-III SPAR Facilities &
Operation
B-I B-II B-III SPAR
Material aspect (cladding &
components)
X X Monitoring;
Wet + Dry -Environment - Components - Fuel assemblies - Workers' dose rate
X X Capacity enhancement - High density racks - Re-racking - Double tiering - Doped coolant - Rod consolidation
X
Degradation mechanisms and models
X X X Fuel conditions - Operational - Fabrication - Technology - Defected fuel rods and assemblies
X X X Changing modes Wet–Dry
X X X
Validation - experimental - experience
X X X Different reactor types
X X X Handling of heavily damaged fuel
X X
System performance
X X X X
3
3. WET STORAGE
Vast quantities of zirconium-alloy-clad fuel now reside in pool storage (145 000 t HM). As of January 1, 2001 by far the largest quantities of spent fuel in wet storage reside in the USA (~42 500 t HM) and Canada (29 600 t HM). the Russian Federation has about 9200 t HM of RBMK-1000 fuel and 3800 t HM of WWER-1000 fuel in storage pools (both reactor types use Zr1Nb-clad fuel). Apart from Canada and Sweden (CLAB), the largest AFR wet storage facilities are generally associated with reprocessing sites: La Hague, Krasnoyarsk, Rokkasho Mura, and Sellafield.
The experience in wet storing spent nuclear fuel now spans some 40+ years. Wet storage continues to dominate as the primary method for storing spent nuclear fuel; ~93% of all spent fuel is wet stored. The benefits provided by this technology are mainly associated with cooling efficiency and shielding. Also, it facilitates safeguards and one-off fuel inspection/examination exercises. Despite the emergence of dry storage technologies, new and extensions to existing wet storage pools have been initiated or completed in France, Germany, Japan, Russian Federation, Sweden, UK and USA.
Reactors have been burning MOX-fuel since the early 1970s. Currently its usage is exclusive to Europe where there are 91 reactors with part-core MOX loading. Spent MOX behaviour in wet storage is similar to UO2 fuel apart from the higher end-of-life radiation levels and associated heat of decay. For these reasons significantly longer wet storage would be required to achieve the comparable radiation & decay heat levels of UO2 spent fuel prior to further processing (interim dry storage, reprocessing or direct disposal).
3.1. SPENT FUEL STORAGE EXPERIENCE 3.1.1. General performance
For primary barrier or containment purposes, cladding corrosion is the factor of most interest in wet storage. However, retention of fuel assembly structure integrity is the more important factor when retrieval is taken into consideration. A detailed review of the degradation mechanisms of various fuel types under wet storage conditions is given in Section 7.
For zirconium alloy clad fuel, data exists for continuous pool storage of greater than 40 years.
This data indicates cladding corrosion to be extremely low (1 × 10-6 µm/a) and, therefore, is not viewed to be the time-limiting factor for prolonged wet spent fuel storage; even in poor pool chemistry conditions.
For stainless steel clad fuels, continuous storage experience of 27 years (LWR) and 23 years (AGR) exists. Although the general cladding corrosion rates for these fuels is significantly higher than zirconium alloy fuel (at ~0.1 µm/a), general cladding corrosion is not a time limiting factor for the storage durations currently envisaged (up to 100 years). For AGR fuel, particular attention to pool water chemistry is required as parts of the fuel stringer become sensitised during reactor operation; see Section 7 and Section 3.2 for further information.
Magnesium alloy clad fuel is particularly susceptible to cladding corrosion under wet storage conditions. Although a protective magnesium hydroxide film is initially formed, the presence of any aggressive ions in the water promotes the dissolution of the protective oxide film and leaves the cladding open to pitting attack. For this reason, Magnox fuel is stored in dosed pool water and storage duration tends to be limited; normally <5 years.
In general the wet storage of spent fuel only appears to be limited by adverse pool chemistry conditions or deterioration of the fuel storage pool structure. Other factors that may influence early fuel removal from NPP pools are removal of long stored fuel to maintain pool capacity, or complete closure of the pool on economic grounds at the time of reactor shutdown. In these cases, one of several options may be considered: interim dry storage, transportation to a centralised pool storage facility, reprocessing of the fuel, or direct disposal (if a route becomes available).
3.1.2. Wet storage experience in the participating countries
The following paragraphs provide examples of the application of wet storage in the countries participating in the CRP.
Canada
Canada has been storing fuel under water since the early 1960s. Fuel has been stored in water pools at CANDU reactor sites for about 40 years.
In 1977/78, AECL and Ontario Hydro initiated an experiment at AECL’s Chalk River Laboratories (CRL) to assess the feasibility of storing irradiated CANDU fuel with zirconium alloy sheath underwater for at least 50 years. One hundred and seventy-six elements from 19 fuel bundles were originally selected for the program. They included prototype generating station fuel irradiated in light water loops in an experimental reactor at CRL and production fuel irradiated in NPD, Douglas Point and Pickering. The oldest bundle selected for the program was discharged from the NPD reactor in May 1962. All 19 bundles were dismantled and the fuel stored as individual elements in open-ended stainless steel tubes. Groups of eight tubes were loaded into either aluminium or stainless steel storage cans. Forty elements were stored in stainless-steel cans in a water pool at NPD; the rest were stored in the NRX water pool at CRL. Stainless steel and aluminium storage cans were selected for the program to investigate differences in storage material behaviour. The NPD pool was selected as one of the storage sites, since its water chemistry and fuel handling techniques were typical of those used in CANDU power stations.
The fuel stored at NPD in the wet storage program was transferred to the NRX pool at CRL in November 1987, when NPD was shut down. Subsequently, in 1994, a decision was made to shut down the NRX pool and as a consequence the wet storage program fuel was transferred to the NRU pool at CRL during February 1995.
The fuel originally selected for the program was characterised in 1977/78, which consisted of a destructive examination of specific fuel elements from each of the bundles plus a non-destructive examination on the remaining elements. The examination revealed no changes in the condition of the fuel or sheath from the original post-irradiation examinations following up to 16 years of storage under water.
In 1988, specific fuel elements from each of the bundles were retrieved for a detailed interim storage examination. That examination revealed no apparent change in the condition of the fuel or sheath in any of the fuel elements without cladding failures following up to 27 years of storage under water. X-ray photoelectron spectroscopy (XPS) examination of fuel fragments from an intentionally defected element that was stored for 21 years under water indicated that the surfaces of the fuel fragments were fully hydrated and significantly oxidised. However, the oxidation was restricted to a thin layer (i.e. nanometers) on the surfaces of the fuel
fragments as indicated by XPS and X-ray diffraction. An interim examination planned for 1998 was not carried out, partially because no observable changes are expected to occur over 10 years. The fuel remains in storage and future examinations will be scheduled as appropriate.
Eastern European Countries (Hungary, Russian Federation)
Apart from one experimental dry storage facility, all spent fuel in the Russian Federation is stored wet either at AR sites (Leningrad, Kursk, Smolensk and Novo-Voronezh NPPs), or at the AFR Mining Chemical Combine (Krasnoyarsk). The behaviour of the Zr1Nb clad WWER and RBMK fuel in storage has been good apart from corrosion in the RBMK spacer grids.
Over ~40 years experience of wet stored fuel exists, there have been no known fuel cladding failures as a result of wet storage; under normal operating conditions.
Corrosion studies have been conducted on Zr1Nb alloy cladding specimen being cut from RMBK and WWER spent fuel assemblies of different burnup and cooling times. Results showed that burnup has a much more significant influence on the cladding plasticity than storage duration.
For RMBK fuel, a maximum oxide layer growth rate of 3–5 µm/year has been observed in the spacer contact area of the spent fuel stored in open canisters with uncontrolled chemistry for 15 years of storage period.
For WWER-440 fuel, the measured oxide film thickness is distributed uniformly along the total cladding; its average value was 0.01 mm. In the location of the spacer grids, a local increase in the oxide film thickness was observed, with an additional value of 0.005 mm [4].
France
In France, because the back-end policy is based on prompt reprocessing, at-reactor spent fuel storage is limited to a short period. After two years maximum of cooling in the NPP pool, spent fuel is transported to the La Hague storage complex prior to reprocessing. The La Hague spent fuel reception, unloading and storage complex consists of the NPH wet cask unloading facility, the NPH storage pool, the T0 dry cask unloading facility and the C, D, E storage pools. The four stainless-steel-lined storage pools are interconnected and offer a total nominal capacity of more than 14 400 t HM.
Spent fuel originating from PWR and BWR NPPs of national and international customers is stored in baskets. This alternative to the usual rack system has been preferred because large quantities of fuel are handled frequently and baskets allow for easier reshuffling. Compact baskets have been designed to handle 9 PWR or 16 BWR fuel assemblies at a time. A storage density of 3 t HM/m2, including all the basket transfer areas for pool operation, is achieved.
Damaged fuel identified during unloading is placed into a sealed bottle prior to storage.
Germany
In Germany, a little less than 2900 t HM spent fuel is presently stored in at-reactor (AR) pools. This number might change a little with time. But it is foreseen that spent fuel will be stored in dry storage casks at the reactor sites (except Obrigheim, where only wet storage is planned), or at the central storage sites in Ahaus, Gorleben or Greifswald. In the future, the amount of spent fuel in AR wet storage will decrease, and finally all fuel will be stored dry.
This means that in Germany the amount of spent fuel stored wet will be limited. Nevertheless, Germany performed a detailed R&D work to assure that spent fuel can be stored wet without problems. The results of the work performed can be summarised shortly as follows.
Assessment of the ability of fuel assemblies to be safely held in wet storage — in particular, for long periods — has been based above all on the evaluation of corrosion mechanisms:
Oxidation and electrochemical corrosion of cladding materials, Corrosive attack of structural components, and
So-called crevice corrosion.
Oxidation is of no importance. Electrochemical attack can be suppressed through the selection of appropriate materials and by adequate control of the pool water chemistry. Corrosive attack on the components is negligible for several reasons: adequate material selection, in-pile passivation, and low temperatures. Crevice corrosion is prevented due to clean water conditions and controlled water chemistry.
All defect mechanisms associated with stress and strain in the Zircaloy claddings can be neglected, since these stresses are much less than the yield strength. A set of spent fuel rods containing incipient flaws as well as through-wall defects penetrating the cladding tube have been stored in a demonstration programme and periodically inspected for 18 years. The inspections have not revealed any unanticipated effects or changes. The demonstration programme provided strong evidence that wet storage of spent Zircaloy-clad fuel in water with adequate water chemistry condition and cleanliness is safe and reliable.
Japan
There are sixteen LWR nuclear power station sites in Japan, and one under construction. All sites have spent fuel storage pools. Approximately half of the cumulative amount of spent LWR fuel discharged from reactor has been shipped to domestic (JNC) and overseas reprocessing plants, and the remainder is stored at reactor sites. Most of these have been stored in pools, without problem.
The total controlled-capacity available for spent fuel storage at all LWR NPP sites is about 14 380 t HM. Due to the controlled capacity being approached at a number of NPPs, and to meet future demands, re-racking (AR) or modification to common-use for two or more reactors has been undertaken. Furthermore, a new pool has been constructed and common use of pools is planned.
Republic of Korea
The integrity of spent fuel rods in AR pool storage has been demonstrated by routine monitoring for over 20 years. No significant problems due to the spent fuel degradation in the pool have been encountered so far. There are no symptoms of fuel cladding or storage component corrosion. However, a top nozzle was separated from the PWR spent fuel assembly during a lifting operation at the Kori-1 spent fuel storage pool. The separation of the top nozzle occurred due to a fracture at the bulge joint area of the sleeve. It was revealed by detailed hot-cell examination that the sleeve failed due to inter-granular stress corrosion cracking. This fuel assembly was handled with a special tool. Recently, the Korea Atomic Energy Research Institute (KAERI) has developed a device for the in-pool reconstituting of sleeve-failed PWR spent fuel assemblies.
Spain
Spent fuel has been wet stored in Spain since 1970. All fuel storage is AR, and storage experience has been good, apart from an occasional top nozzle separation of early manufactured Westinghouse PWR fuel assemblies (see US contribution later). All fuel assemblies in storage have been moved at least once, when the AR pools were re-racked, without any problems occurring.
Sweden
In Sweden, the back end policy is based on final disposal. At the reactor, the spent fuel storage is limited to a short period. Cooling periods of nine months for BWRs, and 12 months for PWRs are required, prior to transport to the centralised storage (CLAB). The transportation of spent fuel is carried out in dry transport casks. The casks are cooled and filled with water before unloading in the unloading section of the CLAB facility. It is expected that CLAB will have to be operational until at least 2050.
The fuel assemblies are transferred from the casks into canisters that contain 25 BWR or 9 PWR FA respectively. When the canister is filled, it is transferred to the underground storage pools.
The current storage capacity is 5000 t HM; however, construction of a second storage cavern with a capacity of 3000 t HM started at the end of 1998 and is due for completion in 2004. As of middle of 2001, the storage records are as follows:
3410 t HM are stored,
Spent fuel maximum burnup is 52 GW·d/t HM, MOX fuel is stored,
Average collective dose is about 100 man·mSv/year,
Release of radioactive nuclides to the environment is negligible, 16 years of successful operation.
United Kingdom
The wet storage of fuel discharged from Magnox and AGR NPPs can be conveniently divided into two phases. The first covers short-term storage at the reactor and the second covers the further interim storage away from reactor at Sellafield, prior to reprocessing. The only exceptions to the above are British Energy’s Sizewell B PWR reactor, where all spent fuel is currently stored at reactor wet, and Magnox Generations Wylfa NPP (see Section 4).
The experience of wet storing commercial spent nuclear fuel on the Sellafield site spans some 45 years. There are four operational AFR storage pools at Sellafield: Fuel Handling Plant (Magnox/AGR), AGR Storage Pond (AGR), LWR Storage Pond (LWR/PIE), and Thorp Receipt & Storage (LWR/AGR). These pools store fuel from the first- and second-generation UK reactors, Magnox and AGR, and also irradiated fuel from LWR reactors, mostly from BNFL’s overseas customers.
The main differences between fuel storage at Sellafield and other facilities are:
Fuel storage is based upon containerised systems (Magnox and AGR containers, Multi- element bottles (LWR)). These facilitate:
• Isolation of the fuel from the bulk pool water,
• The chemistry within the container can be different from the bulk pool water,
• The spread of corrosion products to the storage facility is minimised.
A variety of pool chemistries are utilised depending on the properties of the fuel being stored.
Apart from a few special transports, there are no individual fuel element handling operations, during storage.
The pools are purged with fresh water; resultant effluent is either treated in the Site Ion Exchange Plant, or filtered and sentenced prior to sea discharge.
Fuel integrity/performance under extended storage periods has been monitored by the use of CCTV, activity release models, and full PIE programmes.
United States of America
Spent fuel storage experience from the civil reactor programme spans some 40 years. Apart from the General Electric Morris AFR at Illinois, all wet storage is AR in the USA. In general, the performance of wet storage systems has been excellent with only an occasional operational problem or mishap.
One such event occurred March 2001, when a top nozzle separated from a fuel bundle at the North Anna fuel pool of Dominion Virginia Power. The bundle was being returned to its storage location at the time when it fell some 11 feet back into its designated pool storage slot.
The bundle was a Westinghouse 17 × 17 fuel assembly that had been removed from the core in 1984. The separation occurred at the bulge joint where the top nozzle thimble sleeves connect to the guide thimble tubes. Other similar failures have occurred at Prairie Island (US) and in France, Republic of Korea, and Spain. Stress corrosion cracking has been determined to be the failure mechanism in previous events.
3.2. POOL WATER CHEMISTRY
The role of water in spent pool storage is to:
Facilitate heat removal from the spent fuel, Act as a biological shield,
Maintain fuel cladding integrity, Facilitate spent fuel visual inspection.
To achieve these goals, water quality has to be optimised for the fuel being stored and for the pool components/systems. Additionally, consideration needs to be given to the exclusion of microbiological species (such as algae growth) to prevent loss of water clarity, etc. Factors that may be affected include water chemistry, water treatment, quality control methods and water temperature.
Operating experience gained since the 1950’s shows:
The ingress of aggressive ions such as chloride should be minimised,
Where pool water quality has been maintained, spent fuel storage performance for all types of fuel has been excellent,
Optimum pool water chemistry is linked to the fuel being stored,
Apart from some experience with water-filled, canned RMBK fuel, the general experience with defected fuel is that the defect does not propagate during prolonged storage.
For stainless steel clad AGR fuels, pool water chemistry has to be specified to prevent corrosion of elements 1–5 of the original irradiation stringer through irradiation-induced inter- granular stress corrosion cracking. To inhibit this failure mechanism, AGR fuel is stored in pool water dosed with sodium hydroxide (AFR) and boric acid/sodium hydroxide (AR).
All AGR stations use cooling pools, where spent fuels are stored in borated water. The water chemistry at AGR stations is dictated by criticality control. Boron is added in the form of boric acid. Sodium hydroxide is also added to return the pH of pool water to 7.
AFR pools contain non-borated water and are dosed with sodium hydroxide to maintain a nominal pH of 11.4 (target range: 11.35 to 11.45). Chloride levels of <0.1 ppm are achieved in enclosed pools and 1–2 ppm in open pools. The required chemistry for AGR fuels in the Thorp pools, which do not use sodium hydroxide as corrosion inhibitor, is established when the fuel is placed in containers for storage in the pool (pH ~7, Cl- <0.1 ppm). For experimental purposes, individual containers can be sealed and monitored for activity release.
This method was used to demonstrate that the AGR fuel remained intact during storage in the high quality water in Thorp storage pool even though a corrosion inhibitor was not being used.
For CANDU fuels, which use Zircaloy for cladding material, water chemistry specification is based on a long experience gained over the last 40 years in Canada and 12 years in the Republic of Korea. Spent fuel is stored in AR pools in demineralised water with typical specification given in Table II. Basically, water pH is maintained around 7 while chloride content is limited to a level less than 0.1 mg/kg. (0.2 mg/kg in the Republic of Korea; pH 5.5 to 9.0; and Cl- <0.5 mg/kg at Ontario Hydro, Canada).
The storage of MAGNOX fuels requires particular attention in controlling pool water chemistry. Although magnesium alloy cladding dissolves in pure water by the following reaction:
Mg + 2H2Oo Mg(OH)2 + H2,
a passive/protective magnesium hydroxide film is formed. This protective film, however, is also susceptible to dissolution from acidic species dissolved in the water:
Mg(OH)2 + 2H+l Mg2+ + 2H2O
Table II. Typical examples of water chemistry parameters in storage pools
FUEL TYPE CANDU AGR MAGNOX LWR
Pool type At Reactor pool At Reactor pool AFR pool1 At Reactor pool AFR pool1 CLAB pools Maximum
CLAB pools Normal Storage
structure
Vented containers Ullaged
Container Water Demineralised
water Borated water Demineralised
water Demineralised
water Demineralised water
pH 6.9 7 11.4 11.5 13.02 8.6 5-6
Conductivity 200 µS/m - - 200 µS/m - < 500 µS/m < 100 µS/m Chloride 0.005 mg/kg (ppm) 0.5 ppm < 0.1 to 1–2 ppm < 1 ppm 0.1 ppm < 500 ppb < 5 ppb Sulphate 0.02 ppm 0.5 ppm 0.1 ppm < 0,5 ppm 0.1 ppm < 500 ppb < 5 ppb Fluoride 0.06 ppm - - - - < 500 ppb < 5 ppb
Na 0.1 ppm - - - - < 10 ppb3 < 1 ppb2
99Mo 8.6 × 106 Bq/m3 - - - - -
137Cs 0.3 × 103 Bq/m3 10 × 106 Bq/m3 1 × 107 Bq/m34 60 × 106 Bq/m3
20 × 106 Bq/m3 - 2 × 104Bq/m3
Oil - -
Water activity 9 × 106 Bq/m3 40 × 106 Bq/m3 10.5 × 106 Bq/m3 - 20 × 106 Bq/m3 3 × 106 Bq/m3
1 Purged pools, no in-pool ion exchange systems. Ion exchange if required takes place either on the pool water discharge filter or downstream of the storage pool in a separate facility, prior to sea discharge.
2 pH inside ullaged containers, not pool.
3Na+ and Ca2+
4Pool contains small quantities of fuel with perforated cladding.
11
RBMK PWR and BWR WWER
Pool type AR pool AFR pool AR pool AFR pool AFR pool (UK) 5 AR pool AFR pool Storage
structure Water filled
cans Container (MEB) 6
Water Demineralised water
Demineralised water
Borated water (PWR) Demineralised water
(BWR)
Demineralised water
Demineralised water 7 Borated water Demineralised water
PH 5.5–8 6–7.5 4.5–5.5 4.5–5.5 6.2–7.5 4.3–6.5 5.5–7
Conductivity 300 µS/m 300 µS/m 350 S/m - 200 µS/m 200 µS/m Chloride 0.1 mg/kg 0.05–0.1 ppm 0.15 ppm < 0.1 ppm 0.1 ppm < 0.1 ppm < 0.1 ppm Sulphate - 0.15 ppm < 0.1 ppm 0.1 ppm < 0.150 ppm
Fluoride 0.1 mg/kg 0.15 ppm (Republic of Korea)
< 0.1 ppm -
Na+ and Ca2+ - < 0.5 ppm -
99Mo -
137Cs - 1 × 106 Bq/m3
Oil < 200 ppm Water activity <5 × 106
Bq/m3 <5 × 106 Bq/m3 <2 × 107 Bq/m3
(Republic of Korea) 8 × 106 Bq/m3 2–3 × 106 Bq/m3 40 × 106 Bq/m3 40 × 106 Bq/m3
5 Purged pools, no in-pool ion exchange systems. Ion exchange if required takes place either on the pool water discharge filter or downstream of the storage pool in a separate facility, prior to sea discharge.
6Pool also contains AGR fuel in containers.
12
It is therefore important to minimise ions that would lead to cladding degradation; i.e.
chloride, carbonate, and sulphate. To overcome these effects, Magnox fuel is stored in high purity caustic-dosed water to a minimum pH of 11.4. At AFR, these conditions are increased to pH 13.0 inside the storage containers. Fuel has been satisfactorily stored for up to 5 years at pHt11.4 by dosing with sodium hydroxide and keeping [Cl- + SO42-] <1 mg/kg.
Light water reactor fuels (PWR or BWR)
Apart from a very small amount (three NPPs) of older stainless steel clad fuel, all LWR’s use Zircaloy cladding. Spent fuel is stored in open structures (such as racks or baskets used worldwide), or in sealed containers (such as the Multi Element Bottle used in UK). Water purity is controlled to suppress conditions that might lead to a corrosive environment for spent fuel cladding and structures.
The AR pools for PWR reactors are filled with borated water, while demineralised water is used for BWR reactors as well as for AFR wet storage installations serving both LWR reactor types. Depending on the countries that operate Light Water Reactors, water chemistry for storage installations may be quite different. However, some general parameters can be defined: pH requirement is 4.5–5.5 and chloride and fluoride concentrations have to be limited to 0.1–0.15 ppm (see Table II). In Japan, the pH requirement for BWR stations is 5.3–
7.5; chloride and fluoride concentrations are limited to 0.5 ppm, while each operating value is generally maintained below 0.05 ppm.
For some storage pools that use Boraflex neutron poison material, the increasing concentration of silica may become a concern. The silica results from the degradation of the silicon rubber polymer and the release of other silica that is present as a filler material.
For AR RBMK storage, elements are suspended from the pool floor beams. The fuel has zirconium-1% niobium alloy cladding (Zr1Nb), and the required water chemistry quality has been defined to prevent the corrosion of fuel materials.
In AFR storage installation, RBMK fuels are stored in water-filled cans. Water chemistry analysis in the cans shows it changes with storage time. The simultaneous effects of the absence of water mixing, filtration, and ion exchange lead to an increase in water corrosivity, radiolysis products, and halogen contents. As a result, corrosion becomes more intensive and diversified. The uniform corrosion rate in wet storage, however, is consistent with other zirconium alloy clad fuel. For the major portion of the FA, the corrosion film thickness on the cladding surface was 10–40 µm. However, some FAs developed significant nodules and one FA demonstrated fretting corrosion in the vicinity of the spacer grids. Stainless steel (X18H10T) shows a tendency to pitting corrosion. Alkali additives can be used for suppressing water corrosivity. Spacer grids have now been replaced by zirconium alloy components to overcome the problem.
WWER fuel with Zr1Nb alloy cladding is stored in borated water in the AR pools. Boric acid concentration is maintained within 12–16 g/kg; concentrations of halides (Cl- and F-) are kept below 0.1 µg/kg; pH varies in the 4.3–6.5 range. AFR pools are filled with demineralised water. Halide concentrations are kept below 0.15 mg/kg; pH values are within the 5.5–7.0 range.
3.2.1. Water quality control
The objective of water quality control is to maintain the conditions that (1) minimize corrosion of spent fuel, pool components, radioisotope build-up and inhibit biological growth in the storage water and (2) guarantee the clarity of pool storage water.
To maintain water quality, the following analyses are performed on a routine basis:
PH,
Conductivity, Water turbidity,
Chemical composition (Cl-, F-, SO42-), Isotope-specific activities.
Generally, a complete water activity content is checked weekly, this is increased to a daily analysis in Sellafield’s case (UK) where pool water is discharged to sea after sentencing &
treatment. The complete chemical analysis of the water is generally performed on a monthly basis. At AFRs (France and UK), specific parameters such as chloride, fluoride and sulphate levels are measured either weekly or daily.
Table III provides examples of the common parameters measured, techniques applied and frequency of analysis.
Table III. Water quality parameters
FREQUENCY ANALYSIS METHOD
pH Daily pH electrode
Conductivity Daily Conductivity meter
Turbidity Daily UV-Visible Spectrometer
Cl-, F-, SO42- NO3-
, PO34-
Monthly (AR) Weekly/daily (AFR)
Ion chromatography
Inductive coupled plasma mass spec. (ICPMS)
Alpha, Beta, gamma general control
Weekly (AR) Daily (AFR)
Scintillation counters Ion chamber
Complete chemical analysis Monthly (AR/AFR) On request
ICPMS
Atomic Absorption or Emission Spectroscopy (AAS or AES)
Plasma AES
Where water quality is not strictly controlled or has been allowed to deteriorate, a number of reported incidents have resulted. These include:
Top nozzle separations, etc.
Cladding failures, etc.
3.2.2. Sources of pool water activity
The main sources of pool water activity are from the leaching of corrosion products (commonly known as „crud”) adhered to the fuel, and fuel or fission product species diffusing through any cladding defects and through micro pores within the cladding. Water activity has to be maintained at a reasonably low level to limit dose to the operators and to minimise adsorption by in-pool structural components and ancillary equipment, as a means of facilitating final decommissioning activities. An example at full pool water activity analysis can be seen in Table IV.
Table IV. Example of full pool water activity analysis
Activity, Bq/ml Activity, Bq/ml
110mAg 0.040 698 95Nb 0.105 751
14C 0.034 763 63Ni 2.187 500
144Ce 0.235 838 147Pm 0.004 601
60Co 5.638 653 241Pu 0.420 042
134Cs 0.089 831 103Ru 0.085 452
137Cs 3.937 652 106Ru 0.501 148
152Eu 0.018 595 35S 0.031 119
154Eu 0.007 337 125Sb 0.228 827
155Eu 0.007 929 89Sr 0.099 563
55Fe 0.007 391 90Sr 0.286 458
3H 0.451 346 99Tc 0.085 551
129I 0.019 285 65Zn 0.052 692
54Mn 0.025 567 95Zr 0.156 465
Denotes all results <= Limit of Detection.
Fuel crud
Fuel crud is the common name given to the residues deposited on fuel assemblies as a result of reactions of the coolant with the primary coolant circuit components. It is generally observed that fuel from gas cooled reactors has minor deposits, and PWRs are cleaner than BWRs. This is by virtue of the differences in materials used in construction and the chemistry of the cooling circuits.
In Sweden, crud removed from BWR fuel originating from the Ringhals-1 NPP has been analysed since 1978. This experience shows that the amount of deposited material is more or less linear with the residence time of the fuel in the reactor. The composition of the crud is feedwater water chemistry dependent.
Radionuclidecontent of crud
The radionuclide content of fuel crud mainly comprises of cooling circuit corrosion products that have been neutron-activated. In some cases fission products may be present in trace amounts where in-reactor cladding defects have been present.
Upon fuel discharge from reactor, typically the principal nuclides present are 54Mn,55Fe, 58Co and 60Co, with 55Fe being the dominant species. Due to the longer half-lives of 55Fe and 60Co, only these remain above background levels in 15-year cooled fuel. In the long-term (>100 years), the main activation species remaining is 63Ni, which is present only in trace amount at reactor discharge. An analysis of Ringhal-1 crud is given in Table V.
Table V. Crud composition taken from the fuel cladding at Ringhals-1 NPP in 2000. Average of four (4) fuel assemblies with residence time from 2 to 5 years and burnup from 18.9 to 37.7 MW·d/kg U
Corrosion product g/m2
Co 1.3 E-2
Cr 1.6 E-1
Cu 1.8 E-2
Fe 5.3
Mn 8.3 E-2
Ni 7.6 E-1
Sb 2.3 E-3
Zn 1.9 E-1
Nuclide GBq/m2
58Co 28.7
60Co 52.6
51Cr 15.8
59Fe 8.0
54Mn 13
124Sb 2.8
65Zn 0.6
3.3. INACTIVE CONTENT BWR
Studies indicate BWR crud to be hematite in origin. Typical metal content is ~87 wt% Fe, in the form of red hematite — ~2 wt% Cu, ~4.4 wt% Zn, ~3.3 wt% Ni, and ~2.2 wt% Mn.
PWR
The major components of PWR crud are reported to be spinels of the nickel-substituted ferrite (NixFe3-xO4 with 0 « x d 1) and magnetite (Fe3O4) types. Compounds of NiO, SiO2, and Cr2O3, plus elemental Ni have also been identified. Typical metal content is about 78 wt% Fe, 20 wt% Ni, and 2 wt% other metals (mainly Cr, with Co at 0.03–0.11 wt%).
Gas cooled reactors (Magnox/AGR)
Fuel arisings from gas cooled reactors mainly suffer from carbacious crud deposits from the carbon dioxide coolant reactions with inhibitors/graphite decomposition.
3.3.1. Sources of pool water particulates
The main sources of suspended solids are from particulates associated with the flask if pool discharged, from crud dislodged from individual fuel assemblies, transfers to storage, movement of the fuel, or open fuel baskets in storage. Flask-associated particulates include crud released from the fuel (thermal shock) when water is introduced to the flask internals if shipped dry, dirt on the flask externals, and entrained particulates from the flask liner.
In the case of open storage bays, airborne particulates are also introduced.
3.3.2. Water treatment
Water treatment is necessary to:
Remove suspended solids in order to maintain pool water clarity,
Remove dissolved radioactive species to minimise the dose to operators,
Limit build-up of nutrients (phosphates and nitrates) to minimise biological growth,
Limit build-up of aggressive ions (chlorides, sulphates, etc.) that can initiate corrosion of the fuel assemblies stored, or in-pool components.
Water treatment normally includes a mechanical treatment to remove the solid materials contained in the pool, in conjunction with a chemical treatment to extract both radioactive and non-radioactive chemical species dissolved in the pool water.
Mechanical treatment of the bulk pool water is generally performed by filters (pre-coated sand or mechanical), while chemical treatment is realised with ion exchangers (cationic and anionic resin types are used). In some cases, ion exchange is preceded by neutralisation. Generally only single bed organic ion exchange resins are regenerated when saturated. The resultant concentrate may include a boron recycling step, followed typically by evaporation and final encapsulation for disposal. Mixed bed filters and inorganic exchangers resins are also used. In these cases, there is no regeneration phase and the saturated beds are disposed of directly after encapsulation.
The build-up of particulates on the pool floor and walls is removed mechanically, by in-pool cleaners. A variety of designs have been deployed from modified leisure pool cleaners, simple suction devices to purpose-built two-stage cleaners, coarse and fine (cyclone) filters.
In the case of ion exchange there are three types of systems in operation. These include: out of pool ion exchange columns, ion exchange floated on top of a pre-coated mechanical filter and in-pool water treatment units; examples of the latter include cartridge systems and the combined ion exchange/cooling Nymphea system (France).
Special attention is also required to avoid the growth of microbiological species that can reduce water clarity or even lead to microbial attack of storage materials. The main factors in limiting biological growth are to minimise the introduction of nutrients (especially phosphates), intensity of lighting in the storage area, and temperature.
Where bacterial growth does initiate, treatments have varied form the use of biocides to full- scale mechanical cleaning and collection of the biogrowth. Operating pools at high pH also prevents biological growth.
3.3.3. Water temperature
Water temperature requirements may vary with the specific plant design and with the type of storage installation (AR or AFR). Normal temperature limits are usually based on operating considerations, such as personnel occupation requirements and equipment operating limits rather than fuel corrosion. The temperature limits normally are about 45oC (max 65oC in the Republic of Korea). Other considerations that make it preferable to operate pools at the lowest practicable temperature include lowering the release rate of radionuclides from defective fuels, minimising bacterial or microbial growth, and lowering the humidity level in the storage area.
For AFR storage installations, spent-fuel cooling-time is greater, water temperature is lower and does not exceed 40oC in normal conditions. The influence of water temperature on the transfer of activity from the fuels to the storage water has been shown to double with each 10oC-rise in temperature. It is, therefore, advantageous to keep pool water temperature as low as reasonably possible, thereby delaying release, and taking advantage of the decay of the activity while it is fixed to the fuel surface. The load on the waste treatment systems becomes correspondingly lower [5].
3.4. POOL COMPONENTS AND MATERIALS
The choice of pool components and materials is dependent upon the type of fuel being stored, cost, and to facilitate final pool decommissioning. The latter property has come to the fore as experience in both in-pool performance and final pool decommissioning has been gained.
This has led in most cases to the almost exclusive usage of stainless steel for spent fuel storage structures, heat exchangers, piping and fuel handling equipment.
3.4.1. Pool lining
There are two main methods of treating the internals of the reinforced concrete pool structure to make it watertight, that is to line with welded sheets of stainless steel or to coat with a water resistant paint system.
All LWR AR storage pools are stainless steel lined, while both stainless steel and epoxy liners are used in CANDU AR storage and epoxy/paint is used in Magnox pools. In the case of AFR storage installations all are stainless steel lined (France, Finland, Germany, Japan, Russian Federation, Sweden), with the exception of UK AFR storage pools; where most are only lined at the wind/water line, the rest is painted.
In case of concrete pools coated with epoxy (Canada), the concrete employed has been shown to have negligible corrosive ion leaching and permeability to water. However, when coated concrete is used, the cumulative dose rate on the epoxy has to be limited to prevent epoxy degradation. Measurable changes in epoxy liner properties have been observed after a 1 MGy radiation dose. With the extension of the operating life, this dose limit could be exceeded at some point after the design life of the pool is reached. Since 1988, some extent of radiation-
induced deterioration has been observed at the Pickering station, without any damage on the liner envelope. In the case when water temperature has been maintained higher than 32°C on a regular basis, a risk of minor damage to the coated concrete walls has been observed at the Pickering station in Canada. If epoxy liner radiation-induced deterioration continues, there is a possibility that water may eventually contact the structural concrete. Thus, a programme to investigate the long-term effect of water on concrete was considered.
To resolve the issue of epoxy degradation, alternative paint systems have been developed and deployed in the UK.
3.4.2. Storage racks
The designs/methods used for spent fuel pool storage can be fuel/facility type, and operator dependent. For example, RBMK (AFR) fuel is stored in stainless steel individual bottles suspended in a rack located at the pool surface, in comparison LWR (AR) fuel is typically stored in open top floor mounted borated stainless steel racks. In the case of AFR fuel interim storage, prior to reprocessing, the storage systems are designed to facilitate the bulk movement of fuel and as such tend to have a reduced capacity than can be achieved AR by the fixed high-density rack systems. In one country (UK), the storage systems mainly comprise of the „basket” used in transportation; i.e. the Magnox & AGR skips and the multi-element bottle (LWR).
The materials of construction are primarily either borated stainless steel, or stainless steel, usually in combination with a neutron absorbing material; for example Boral, Boraflex or Cadminox. One exception is the Magnox skip, painted mild steel, which is used due to concerns with respect to the potential for electro-coupling between the fuel cladding and stainless steel. The only other material that has found some use for structural components is aluminium; mainly for its corrosion resistance in deionised water when passivated.
4. DRY STORAGE 4.1. INTRODUCTION
Almost 20 years of favourable experience exists with the dry storage of spent power reactor fuel and about 30 years with research reactor fuel. Dry storage experience exists with fuel from a variety of reactor types (CANDU, HWR, PWR, BWR, WWER-440, WWER-1000, RBMK, MAGNOX and the HTGR). Since its conception, dry storage of spent fuel has evolved into a wide variety of systems. Examples of these are concrete canisters, steel-lined concrete containers and concrete CANSTOR modules (modular vault-like storage system) in Canada, concrete canisters in the Republic of Korea, vaults in France, Hungary, UK and the USA, and casks in Germany, Japan, Russian Federation (in development), Spain and the USA [1, 2, 3, 6]. At the present time, all countries participating in SPAR (Canada, France, Germany, Hungary, Japan, Republic of Korea, Russian Federation, Spain, United Kingdom and the USA) are engaged in various dry storage technologies. Almost all participating countries are also actively pursuing a dry storage research and development programme. So far, the results of the research indicate that fuel can be stored safely under the present conditions for many decades.
As regards licensing conditions for the dry storage facilities, different trends and licensing periods have been implemented throughout the SPAR countries as shown in the following table VI:
Table VI. Dry storage licensing conditions in the SPAR countries
Country Initial licence period Renewal period
Canada 2 years 2 years
Germany 40 years N/A
Hungary 10 years 10 years
Japan Not limited (similar to the
reactor site licence)
N/A Republic of Korea Not limited (covered under
the reactor site licence) N/A
Russian Federation 1 year 3 years
Spain 20 years 20 years
United Kingdom Not limited (covered under
the reactor site licence) N/A
USA 20 years To be determined
N/A = not applicable.
Dry storage has become a mature technology and the quantities being placed into dry storage are increasing significantly. The inventory of spent fuel in dry storage in the above countries, as of 1 January 2001, was about 8693 t HM. The largest quantity is being stored in Canada, where since 1996, about 25 000 spent CANDU fuel bundles are being placed in dry storage annually. This is expected to increase when the Bruce Dry Storage Facility comes into operation by the end of 2002. At that point the OPG facilities will start placing in dry storage about 76 800 bundles per year. In Canada, the total inventory of spent fuel in dry storage at the end of 2000 was 4239 t HM. Also, in the USA, as many of the spent fuel storage pools reach their capacity, dry storage is a significant factor in the utilities` spent fuel storage strategy.
A helium storage environment is used in most of the systems, although air as well as helium is used in Canada. In the Republic of Korea only air is used to store CANDU fuel. Air is also being used for the experimental dry cask storage of RBMK-1000 fuel in the Russian Federation and a combination of CO2 and air in the vault for the dry storage of MAGNOX fuel in the UK. Nitrogen is used in the MVDS in Hungary and at Ft. St. Vrain (USA).
Dry storage of spent LWR fuel in an inert atmosphere is licensed dependent on burnup and type of cask for temperatures up to 410oC in Germany and 380oC in the USA and Spain [7, 8].
Dry storage in nitrogen is licensed for a temperature of 350oC in Hungary for Zr/Nb clad fuel.
Here, an application is being discussed with the regulatory body to allow raising this value, to at least 380oC. Dry storage is licensed in Canada and the Republic of Korea for temperatures up to 160oC for storage in air. In Canada, at the Pickering Station, storage is also licensed for temperatures of up to 360°C in a helium atmosphere.
In most cases, the dry storage containers are loaded with spent fuel under water. Following removal of the bulk water, the container requires vacuum or hot gas drying to prevent later aqueous corrosion and hydriding of the Zircaloy components and cask construction materials.
Care must be taken to ensure that the vacuum cycle is slow enough to allow full evacuation of all the water in the cask and also to prevent freezing of residual water due to evaporative cooling within the cask. Additionally, the drying level specification for reaching adequate dryness levels must be carefully defined. The drying processes utilised today are vacuum and hot-gas drying.
The vacuum drying process involves lowering the cover gas system pressure below the vapour pressure of the water at the drying temperature. The container is considered to be satisfactorily dry when the system pressure (e.g. ~6 to 10 millibar) remains constant for a specific time period (e.g. ~20 minutes). However, large amounts of residual water or relatively cold fuel may require a fairly long drying time or more sophisticated methods. The evaporation rate will also depend upon the temperature of the surrounding area. Vacuum drying is utilised in Canada, Germany, Spain and the USA.
In the hot-gas drying process, (process temperature 90–150oC), the fuel assembly is uniformly heated by the hot gas, to evaporate the water, in addition to blowing it away. This ensures complete removal of water residuals from all parts of the fuel assembly. The process time depends upon the heat input and flow rate of the hot gas. This process is utilized in Canada, Hungary, and the Republic of Korea. The fuel assembly container is considered to be dry, when the moisture content of the exhaust air equals the moisture content of the inlet air.
Most of the fuel in dry storage is clad with a zirconium alloy (Zry-2, Zry-4, Zr2.5Nb and Zr1Nb); however, dry storage experience also exists for magnesium and aluminium clad fuels. Average burnups of spent fuel presently in dry storage range from 4.5 to 33.5 GW·d/t HM, while the maximum burnups range from 7.5 to 50 GW·d/t HM. However, there is an almost universal tendency towards increasing the discharge burnup of the fuel elements. In Germany, for instance, average discharge burnups for PWR fuel assemblies have increased from 35 GW·d/t HM in 1983 to 50 GW·d/t HM in 1998, and a value of 65 GW·d/t HM is expected by 2003 in this decade. In the USA and other SPAR countries, most of the spent fuel assemblies are being, or will be in the very near future, discharged with burnup in excess of 45 GW·d/t HM. However, most presently licensed storage systems have burnup limitations of 45 GW·d/t HM, or less.
4.2. DRY STORAGE STATUS AND EXPERIENCE
Dry storage experience in the countries participating in SPAR CRP is described in this Section. Since the inception of the previous BEFAST project, other countries have also developed dry storage systems that are not reported here. As the amount of spent fuel stored dry in the SPAR countries is a large percentage of the total amount stored in dry conditions, this experience is considered to be representative of the total population. Total amounts of SF dry stored as of January 1, 2001 in the SPAR countries are shown in Table VII.
Canada — Atomic Energy of Canada Limited (AECL) has been storing spent fuel from its research reactors in a dry environment at the Chalk River Laboratories (CRL) for several decades. AECL has also designed reinforced concrete canisters (CCs) which have been used since the mid 1970s to store fuel from several NPPs. AECL has also developed two modular dry storage systems, CANSTOR (for CANDU fuel) [9] and MACSTOR (for LWR fuel) [10].
These are reinforced concrete modular structures that store spent fuel inside metal containers.
A CANSTOR system was licensed in 1995 for the storage of spent CANDU fuel from Hydro Quebec's Gentilly-2 Nuclear Generating Station (NGS). Loading of the first CANSTOR module began in the latter part of 1995.
Table VII. Amount of spent fuel in dry storage
Country t HM in dry storage
(As of January 1, 2001) t HM in dry storage (Forecast by 2010)
Canada 4 239 16 200
France 50 50
Germany 161 1700
Hungary 224 800
Japan 75 *
Republic of Korea 914 3 400
Russian Federation 0 6 000
Spain 0 220
UK 650 0
USA 2 430 ~18 000
* The dry storage facilities at the Fukushima-Daiichi and the Tokai-Daini stations have capacities of 150 t HM and 260 t HM respectively. The amount of SF in dry storage in 2010, however, is not clear at this time.
Ontario Power Generation (OPG) uses Dry Storage Containers (DSCs) of its own design.
They are transportable containers built of reinforced concrete, encased in (inner and outer) steel liners, which each hold 384 CANDU fuel bundles. OPG has had one operating dry storage facility at the Pickering site since 1995, and in January 2000 obtained a construction licence for a new storage facility at the Bruce nuclear site. The Bruce facility is expected to come into service by the end of 2002. The combined capacity of both facilities is 1.23 million bundles, and will provide sufficient storage for the rest of the stations’ operating life.
Canada has not selected yet a strategy for long-term management of its spent fuel. This decision will be made under the framework of new nuclear waste legislation, which is expected to be enacted in 2002. Under this law the Canadian nuclear waste (spent fuel) producers will be required to create a separate waste management organisation (WMO) with the mission of develop and implement an integrated strategy for the long term management of all the spent nuclear fuel generated in Canada. The terms of this nuclear waste legislation are dictating research priorities in 2002. Current research being supported by the nuclear waste owners in Canada is focused on the requirements of the new law, the primary requirement being to issue a report examining options for the long-term management of nuclear fuel and recommending an approach. That report must be issued within three years of the law being enacted. Subsequently, the government will use the results of the study to select a preferred approach for management of the spent nuclear fuel, for which the WMO will become the implementing organisation.
France — Dry storage has been developed in France at the CASCAD facility for spent fuel that does not require prompt reprocessing. As of 1 January, 2001, approximately 50 t HM of heavy-water reactor fuel from the decommissioned EL4 Brennilis reactor had been stored in the vault at the CASCAD facility at Cadarache. The maximum capacity of the facility is about 100 t HM. Spent fuel is canisterized at the reactor (within a dry cell). Canisters, vacuum dried and filled with helium, are transferred to the dry-storage installation. Canisters are stored in storage wells ventilated by natural convection.
